＂Tritium Well Depth＂ and ＂Tritium Well Time＂

Somewhat similar to, but quite different from the xenon-135 poisoning effects resulted from fission produced iodine-135 via β-decay during restart-up process of a fission reactor, a complete new concept of ＂tritium well depth and tritium well time＂ is first time introduced in fusion research area by authors. It shows the least required amount of tritium storage is to start up a fusion reactor ,and the least operating time for achieving the ＂tritium break even＂ during the initial start-up phase due to the finite tritium breeding time. The tritium well depth and tritium well time depend on the tritium recovery scheme and extraction process, the tritium retention of reactor components,     

Research and Development of SWITRIM Code and Its Applications

The ftrst numerical simulation code package WITRIM has been developed to calculate the tritium inventory distribution and time-evolution in all sub-systems of FEB fusion reactor. The applications during recent six years indicate that it is reasonable and fully admitted by colleagues abroad. Some creative papers with new concept are published. For instance, we first time pointed out a new phenomenon of ＂tritium well depth and tritium well time＂ during the fusion reactor start-up phase. This is somewhat similar to, but quite different from the ＂iodine well depth and iodine well time＂ poisoning problem during restart-up process of a fission reactor. The authors not only proposed but also numerically solved this new phenomenon. The combination of the SWITRIM code package, user＇s guide, and application example are briefly introduced in this article.     

New Mechanism for Enhancing Ash Removal Efficiency and Reducing Tritium Inventory

A new mechanism is suggested to suppress ash particle back streams in the divertor region of our fusion experimental breeder （FEB） reactor for enhancing the ash removal efficiency and reducing the tritium inventory by applications of the nonlinear effect of high power rf ponderomotive force potential which reflects the platereleased and re-ionized He^＋ back to the plate. Meanwhile, the potential does not hinder α particles, which are coming from scraping of the layer, flowing to the target plate. However, it does stop tritium ions flowing to the target. Based on the FEB design parameters, our calculations have shown that the ash removal efficiency can be improved by as much as 40% if the parallel component of rf field 150-200 V/cm is applied to the location at a perpendicular distance L = 20 cm apart from the plate and the plate-recycling neutral helium atom energy is＇ about 0.75eV, at the same time, the tritium inventory can be reduced to some extent.     

An Ingenious Approach of Determining Hydrogen Isotope Solubilities, Diffusivities and Permeabilities in GWHER-1 Stainless Steel

To predict the total tritium inventory for starting up a fusion reactor and the tritium distribution in all components of reactor system including test blanket module （TBM）, the hydrogen isotope solubilities, diffusivities and permeabilities are urgently required. Moreover, the neutral hydrogen isotopes released from the plasma facing component materials and their retention in blanket structural materials have notable impacts on either neutral particle transport, recycling in the edge plasma, and density profile control during discharge in present devices, or tritium extraction and recovery in future tokamak reactors. The experimental investigations of hydrogen isotope diffusivities and solubilities in GWHER-1 steel （hydrogen- embrittlement-resistant stainless steel made by China-great-wall-steel-mill） have been performed.     

Neutronics Calculation of ITER HC-SB TBM

ITER＇s test blanket modules （ TBM ） is a test-bed to demonstrate tritium self-sufficiency and extraction of high-grade heat for a future fusion reactor. It is also a test plateform to test electro- magnetic, thermo-hydraulic and tritium breeder for DEMO blanket relevant technologies. A great deal of the largest and the most important nuclear issues are related to neutronics. In consideration of strict requirements of absolute safe operation for ITER and TBM, all of probable or potential problems of TBM must be investigated such as power generation, tritium generation, thermo-hydraulics and energy production and so on.     

Measurement of Neutron Fluxes and Spectra during the NT-TBM Operation of CN HCSB

Neutronics measurement system provides neutron fluxes and spectra at the locations of beryllium multiplier and tritium breeder during the operation of the NT-TBM （D-D and D-T phase in ITER）. This is important for studying and assessing capabilities of beryllium multiplier and tritium breeder in the China helium cooled solid breeder （CNHCSB）. A special neutron diagnostic system has been proposed that allows to measure neutron fluxes and spectra without interrupt the operation of ITER. This system includes encapsulated foil activation analysis, micro-fission chamber detectors （MFC）, and a compact neutron spectrometer using natural diamond detector （NDD）.     

Study on Chemical Behavior of Energetic Deuterium-implanted into Li2TiO3

In ITER test blanket module design, the elucidation of tritium recovery from solid tritium breeding material is one of critical issues and Li2TiO3 is a candidate of tritium breeding materials. In the present study, in order to understand tritium behavior in solid breeding material Li2TiO3, the X-ray photoelectron spectroscopy （XPS） and thermal desorption spectroscopy （TDS） experiments are carried out to elucidate the trapping behavior using 3 keV D2^＋ implanted sample. The Ti-2p XPS spectrum shows that a shoulder appeared at the lower energy side as increasing ion fluence, and it was suggested that Ti^3＋ was formed by the reduction of Ti^4＋.     

Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB) to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDSFBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW.yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.     

A Newly Developed Non-destructive Tritium Measurement Technique and Its Application to a V-4Cr-4Ti Alloy

Several evaluation techniques of tritium in material have been developed so far, such as imaging plate method or electrochemical etching and thermal desorption analysis, as well as conventional β-ray counting. For the latter, its detectable depth is usually limited to a very thin surface region, for example, about 1μm depth for organic and 0.1 μm for metallic materials owing to the shallow escape depth of tritium β-rays.     

Preliminary Design of ITER Test Blanket Module HC-SB Structure

The helium cooled solid breeder blanket that represents the main stream in relevant blanket concepts. It is considered as one of the main options for the fusion reactor. This concept have advantage as follows： （ 1 ） structure simpleness and stability, （ 2 ） no effect of MHD, （ 3 ） a good compatibility in different materials.     

FEB—E氚循环系统的计算机模拟

对聚变实验增殖堆（ＦＥＢ）工程概要设计的氚燃料循环构造了一个动态子系统模型，研制了模拟氚燃料循环系统的计算机程序ＳＷＩＴＲＩＭ，计算运行一年后１０个子系统中的氚投料量和整个推系统总的氚投料量，这对预示起动一个聚变热功率的１５０ＭＷ量级的实验增殖堆所需的最少初始氚投料量有参考价值，计算结果表明，要求的最少初氚贮备量除了与燃料气体净化系统和同位互分离系统中氚的平均逗留时间有关外，还与包层液态锂中提取氚     

中国聚变工程实验堆启动氚投料量与氚平衡分析

根据中国聚变工程实验堆(CFETR)堆芯设计参数及燃料系统流程模型，采用平均停留时间方法，建立燃料循环子系统的氚转移模型用来描述氚在各子系统之间的输运、滞留等过程。采用该模型，分析了不同聚变功率水平、运行因子以及燃烧率对中国聚变工程实验堆的氚平衡以及启动氚投料量的影响。     

SWITRIM��������ƺ�ʹ�øſ�

研制出了用于计算氚投料量在FEB聚变堆各个子系统中的分布及其随时间变化的数值模拟程序包SWITRIM。通过近5年的使用，表明其运行良好、计算结果可靠。用SWITRIM数值模拟研究了聚变堆起动过程中的“氚坑深度和氚坑时间”新现象。简单介绍了SWITRIM程序包的组成和用户使用说明以及最新的运用等。     

聚变堆氚系统设计中的一些重要问题研究(Ⅱ)

在文献[1]中,计算了FEB-E 聚变堆PFC 材料内的氚滞留量、堆系统总的氚投料量、启动运行开始阶段的氚坑深度和氚坑时间大小。这里将讨论在ITER 的TBM 氚增殖包层内固体氚增殖剂中的氚如何高效率地被载氚气体带出并且以高效率地提取回收。本部分将进行创新的探索性研究并且提出某些减少氚滞留量和改善氚提取回收效率的新方案,例如：基于氘饱和的海绵效应;第一壁表面建立氘和铍的伴同沉积层;基于在低频外电场作用下载氚气分子和硅酸锂颗粒电极化旋转催化同位素交换速率的增强载氚气提取氚效率“SPB 方法”。     

�۱���ϵͳ����е�һЩ��Ҫ�����о�(��)

In part one published in the last issue, the tritium retention and the total tritium inventory in PFC materials of FEB-E fusion reactor had been calculated. The tritium well depth, tritium well time during the FEB-E fusion reactor start-up and initial operation phase had been obtained. In this part, how to improve tritium recovery efficiency in the ITER TBM solid breeder blanket with using purge gas has been discussed. Some new innovative schemes for reducing tritium retention and improving tritium recovery efficiency are proposed. Such as, sponge mechanism based on deuterium saturated PFC materials; deuterium and beryllium co-deposition layer created on first wall surface; SPB scheme for enhancing tritium recovery efficiency of purge gas in ceramic breeder blanket based on the electrical polarization rotations catalyzing isotope exchange rate enhancement resulted from applied low frequency electric-field, of Li4SiO4 grain and purge gas molecular particles and so on, are explored.     

锂铅合金释氚实验研究

由于锂铅合金因具有高增殖比、低活泼性和可能作为冷却剂的特点,被认为是最有潜力的能源堆包层氚增殖材料。在理论模型描述熔融锂铅合金氚释放行为的基础上,开展了中子辐照后Li₁₇Pb₈₃合金的离线氚释放实验。结果表明: 释放氚的化学形式99％以上为难溶于水的成分（HT或T2）; 氚滞留时间随载气中氢分压的增加而减小,氢分压达到1000 Pa后变为常数,且与实体积无关;氚释放速率对温度的依赖性符合Arrhenius定律。以此为基础得到的氚在熔融锂铅中的动力学参数结果,虽与文献值有差异,但同样证明了在633—973 K的范围内, 氚从液态锂铅到气相的整个释放过程中起决定作用的是氚在合金内的扩散和气\|液界面的多相反应重组。Lithium\|lead alloy is considered to be one of the most prominent tritium breeding materials for the fusion reactor blanket because of its high breeding ratio, and low reactivity and possible use as coolant. An out\|of\|pile experiment of tritium release from Li₁₇Pb₈₃ alloy was performed after neutron irradiation on the base of mathematical model to describe tritium release behavior from an eutectic lithium\|lead alloy. The results suggest that the dominant chemical form of the released tritium (>99%) was the water\| insoluble component (HT or T2). Tritium residence time decreased with increasing H₂pressure in carrier gas up to 1000 Pa, and above this concentration limit it became constant and not influenced by the plenum volume. The temperature dependence of the tritium release rate can be described by an Arrhenius law. Consequently, the present results on the kinetic parameters of tritium in molten Li₁₇Pb₈₃alloy are considered to be different from the values in literature, but it is the same that the overall release process is governed by the diffusion of tritium atoms in the Li₁₇Pb₈₃and by the heterogeneous reaction at the gas\|eutectic interface of the tritium atom recombination at temperatures from 633 to 973 K.     

�۱���ϵͳ����е�һЩ��Ҫ�����о�(��)

聚变堆第一壁表面和PFC材料内的氚滞留量、堆系统总的氚投料量多高？在启动和运行的开始阶段的氚坑深度,氚坑时间的大小是多少？在TBM氚增殖包层内固体氚增殖剂中的氚能否高效率地被载氚气体带出来并且以高效率地提取回收？能否找到某些新机制解决这些问题是决定实现ITER的预期目标和最终实现聚变能的实际运用成败的关键问题。本文第(Ⅰ)部分回答前面两个问题,在下期第(Ⅱ)部分将进行创新的探索性研究并且提出某些减少氚滞留量和改善氚提取回收效率的新方案,例如：基于氘饱和的海绵效应;第一壁表面建立氘和铍的伴同沉积层;基于在低频外电场作用下载氚气分子和硅酸锂颗粒电极化旋转催化同位素交换速率增强提高载氚气提取氚效率“SPB方法”等等。     

ˮ���մ���ֳ����������˷���

Based on the design of the 2015 version of China Fusion Engineering Test Reactor (CFETR) water cooled ceramic breeder (WCCB) blanket modules surrounding the plasma, a tritium transport model has been developed. Tritium transport analysis has been carried out for each blanket module with different breeding zones, purge gas loop, coolant loop and steam generator. The results indicate that the concentration, permeability and retention of tritium among blanket modules are different. For all of the WCCB blanket modules in CFETR, the tritium retention inside the breeder is 6.62×10^-2g, the tritium retention inside the structural materials is 2.01g, the tritium retention inside purge gas and coolant loop are 4.03×10^-4g and 0.19g respectively, the tritium permeation through the steam generator tube walls is 20mg•y^-1, the tritium permeation from the coolant pipes is 0.1mg•y^-1.     

增殖剂球床载气体流动特性

增殖剂球床是聚变堆或混合堆产氚包层可选结构之一,准确把握增殖剂球床中载带气体的流动特性有助于提高对球床载氚过程的认识并优化包层设计。采用离散元程序PFC3D模拟增殖剂小球的填充行为,在球床内不同位置随机截取不同尺寸的控制体,利用布尔运算中的"差集"得到孔隙范围,建立孔隙分布的三维几何模型,进一步划分网格并用计算流体力学(CFD)方法求解,得出控制体上单位长度的压降以及单元体内的速度分布特征,计算结果发现载带气体速度分布与γ分布很类似,且只要选取恰当的控制体,通过计算流体力学方法可以较好地分析整个球床孔隙内流体的流动,有利于进一步研究载氚及相关过程。     

CFETR ˮ���մɰ���ʱ����ز�����Գ�������

水冷陶瓷包层是中国聚变工程实验堆(CFETR)的三种候选包层概念之一。基于CFETR水冷陶瓷包层的一维中子学模型,通过蒙特卡罗输运模拟程序MCNP和活化计算程序FISPACT的耦合计算,经三维转换系数修正,分析了CFETR水冷陶瓷包层时间相关产氚特性。结果表明,当CFETR运行因子为0.5,聚变功率为200MW时,水冷陶瓷包层在运行5年、10年、20年后,氚增殖率(TBR)的降低都不显著,但是年产氚剩余量的降低很明显。此外,产氚包层内初始时刻TBR对产氚特性的影响也很大。